L-81-462, Forwards Responses to 810821 Questions 1,2 & 5 Re Pressurized Thermal Shock to Reactor Pressure Vessel

Forwards Responses to 810821 Questions 1,2 & 5 Re Pressurized Thermal Shock to Reactor Pressure Vessel

ML17341A590
Person / Time
Site:	Turkey Point
Issue date:	10/21/1981
From:	Robert E. Uhrig FLORIDA POWER & LIGHT CO.
To:	Eisenhut D Office of Nuclear Reactor Regulation
References
L-81-462, NUDOCS 8110270244
Download: ML17341A590 (10)

Text

RKGULAiTORY NFORMAiTIOi4ISTRIBUT'ION SYb EH>> (RIDS)

SUBJFCiT:: For war ds responses to 810821 Questions 1 i2 8

5>> rei pressurized the'rmal shack to reactor pressure'essels::

DISTRIBUTION CODEl:

A049S COPIFS RECKKYEDtLiTR ENCL'IITLKI:i Thermal Shock to Reactori Vesseil SIZE>>:: '

NOTES:"

ACCESSION NSR:81 f 0270244 DOC ~ DATE(! 8'1/10/21 NOT'ARIZED:

NO DOCKEiT'nt FACILi;50.-250 Turkey Point Pl anti Uni t" 3i Florida Power and I ign't C'50CQQ~S 50'. 251 Turkey Point., Plant'p Unit>> 4< Florida Power and Ligplt C

05000251 AUTH~,NAMKl AU1'HOR'FF ILI'ATiION UHRI 9 i R ~ K' Florida Power li'ight" C'os RE C'I P ~ 4AM-"I RECiIP'IENT AFFIliIAiTIONi EIISE VHUT'jDG ~

Divisi on of Licensing.

RECIPIENT IDi CODE/NAMKI ACT~ION::

OR'Si 41 BC(

01 ~

1?!

07 INTERNAL@ A/Di SA'/DLl AEOD BASDEKASgD DIRECTOR i DE)

DIRECiTQR p DLi DIRECTOR:i DST ELOi I SKI JOHNSOV'i C.'iIAWi 8

VRR'/DHFS DEPY09" OR" ASSESS BR 11 R

<LECKER REAC SYS BR RK8 DEITI YAGINS~N 8

rl'AZELiTON COPIKB LTITR EiVCLi 13 13 1

1 1

1 1

1 1

1 1

1-1 1

0 2'.

2 1

1 1

1 1

1 0

1 1

1

.1 1

1 1

1 1

1 RECIPIENT'Di CODE>>/NAME ABBOiTT,g K'USTiINiJ GROUSE.gR O'IRECTOR'r DHFS O'IRKCTOR i DSI DiIRECTORiNRR GOOD'WIN't E" IGNE>rI JOHN SOiV'g R'iARSHiTf NRR'/DL DIR PROC/TST REV RA BR G FILE(

VISS ING e G ~

'AOODSiRl 05 04 COPIKB LAiTRl KNCLi 1

1 1

1 ~

1 1

1

.1>>

1 1

1 1'

f>>

1 1>>

1 1>>

1 1-1 1

1 1>>

1 1>>

1 1

1 KXTERNALi: ACRS iVRCl PDR<

iVT'I S 10.

0>>il 16 16 1,1 1

1 LPDR'VS I Ci 03 06' 1

1 1

TOTAL NUiiBERi OF COPIES REQUIRED 'iTTR 68 ENCL(

66

FLORIDA POWER II LIGHTCOMPANY October 21, 1981 L-81-462 CQ Office of Nuclear Reactor Regul ati on rS Attention:

air. Darrell G. Eisenhut, Director

/'rggfI/@

Division of Licensing U.S.

Nuclear Regulatory Commission 0Cyp 8 )

Washington, D. C.

20555 8 188)~

,'ear Mr. Ei senhut:

Re:

Turkey Point Units 3 8

4 Docket Nos.

50-250 and 50-251 Pressurized Thermal Shock to Reactor Pressure Vessels Please find attached

-our response to guestions (1), (2) and (5) of your letter dated August 21, 1981.

Our letter L-81-418 dated September 23, 1981 outlined our schedule for responding to the remainder of questions, and described our participation in the Westinghouse Owner's Group.

The results of that effort may be used to supplement our response to question (5) in the future.

Very truly yours, Robert E. Uhrig Vice President Advanced Systems and Technology REU/JEM/mbd Attachment cc:

Mr. James P. O'Reilly, Region II Mr. Harold F. Reis, Equire PDR 0244 8i-102-iW 8110@j

<<>000ai0

'DR.

PEOPLE... SERVING PEOPLE

Pressurized Thermal Shock to Reactor Pressure Vessels Provide the RTNDT Values of the critical welds and plates (or forgings) in your vessel for:

(a)

Initial (as built) conditions and location (e.g.

1/4T) and (b) current conditions (include fluence level) at the RPV i nside carbon steel surface.

Res onse 1:

Initial Mater i al

~RT~

Current h~RT ~

~RT~

(a)

Intermedi ate Inner 123P481VA1

+50 F

Forging*

+35 F

+ 85F Ci rcumferenti al'Gi rth) i<el d**

Lower Forging*

SA 1101 122S180VA1

+3 F

+40 F

+190 F

+193F

+35F

+ 75F 1/4 T

Inner wall.

The current RTNnT (1/4 T)

= +168 F.

Value is based on Unit 3 data which has been shown to be more representative of Unit 4 than surveillance capsule removed from Unit 4 (L-77-113, dated April 11, 1977 and L-77-326, dated October 21, 1977).

+

Based on the slo e of prediction curves presented in proposed ASTM Standards "Pre icting Neutron Radiation Damage To Reactor Vessel Materi al."

(b)

There have been 5.61 Effective Full Power Years (EFPY) of operation as of September 30, 1981 at Turkey Point Unit 4.

The total fluence on the inner wall is 1.1 X 10 n/cm and 6.6 X 10 n/cm2 at 1/4 T.

guesti on 2:

At what rate is RTNDT increasing for these welds and plate material?

Res onse 2

RTNDT is increasing at the rate of 7'/F/EFPY for the next 10 years; for the remainder of life, 5'/F/EFPY.

The rate of change for the forgings is 30$ for the remaining design life of the vessel.

These are based on, the ~slo e of prediction curves presented in proposed ASTM Standards "Predicting Neutron Radiation Damage To Reactor Vessel Material."

uestion 5:

Provide a listing of operator actions which are required for your plant to prevent pressurized thermal shock and to ensure vessel integrity.

Include a

description of the circumstances in which these operator actions are required to be taken.

Included in this summary should be the specific pressure, temperature and level values for: a) high pressure injection (HPI) termination criteria presently used at your facility, b)

HPI throttling criteria and instruction presently used at your facility and c) criteria for throttling feedwater presently used at your facility.

For each required operator action, give the information available to the operator and the time available for his deci si on and the requi red acti on.

State how each requi red operator acti on i s incorporated in plant operating procedures and in training and requalification training programs.

RESPONSE

5 Previous

analyses, as well as more recent work by the Westinghouse Owner's Group have confirmed, using conservative assumptions and methods, that a

potential reactor vessel integrity problem is not an immediate concern.

However, additional detailed analyses of reactor vessel integrity and investigation of remedial actions are also being performed and will be available at the end of this year.

For the low probability accident events that pose a concern, the following are a

summary of criteria for operator action relevent to mitigating these events.

Analyses performed for the LOCA requires no operator action.

The operator action required for the Large Main Steam 'Line Break is assumed to occur within 10 minutes and we expect this to be a reasonable assumption.

MAIN STEAM LINE BREAK ACCIDENT Criteria for Hi h Pressure Safet Injection Termination:

The Emergency Operating Procedure for the loss of Secondary Coolant Accident instructs the operators to terminate the HPI if the following conditions are sati sfi ed:

Conditions for~earl termination of high pressure (1450 psig) safety injection:

If one wide range reactor coolant temperature TH (as confirmed by Core Exit Thermocouples) is less than 350'F and pressurizer level is above 20K, span and reactor coolant is above 700-psig and subcooling (margin) above 60'F and one or more steam generators narrow range level above 1@, span or Auxiliary Feedwater Flow of 570 GPM

Criteria for Throttlin Auxiliar Feedwater:

When faulted steam generator is identified, isolate auxiliary feedwater flow to faulted steam generator.

When any,intact steam generator narrow range level is in the narrow range, control steam generator level by throttling auxiliary feedwater flow.

Criteria for Throttlin Hi h Pressure Safet In ection:

None LOSS OF COOLANT ACCIDENT Criteria for Hi h Pressure Safet In'ection Safet In ection Termination The Emergency Plant Operating procedures instruct the operators to terminate the HPI in case of LOCA under the following conditions:

Pressurizer Level is above 50$ span and Reactor Coolant Pressure above 2000 psig and increasing and Subcooling (margin) above 30'F and Auxiliary Feedwater Flow of at least 570 gpm or steam generator level greater than 15$ narrow range span in one or more steam generators Criteria for Throttlin Auxiliar Feedwater Only after steam generator water level is established on; the narrow range should Auxiliary Feedwater System Flow be reduced to maintain required level.

Criteria for Throttlin Hi h Presure Safet Injection None The shutoff head of HPI pumps is 1450 psig, thus limiting, the repressurization of the reactor vessel.

These actions are taken primarily to assure that under the most adverse conditions, core cooling is maintained thus assuring vessel integrity.

Our NSSS vendor has not identified any operational transients that pose a threat to the reactor vessel integrity.

The low probability accident scenarios mentioned above pose a concern and will be addressed'.

As noted above we are participating in the Westinghouse Owner's Group and will modify our procedures based on their results as appropriate.

The following: instruments are available in the Control Rocm to allow the operator to monitor the course of accident and take appropriate action:

REACTOR COOLANT. SYSTEM V

PA-3'PA-3 Wide Ran e

0-700F Tem erature Recorders TemPerature Hot Leg (TH), A, B, and C

Temperature Cold Leg (TC)

A B

and' Narrow Ran e

.540 - 610F Tem erature Recorder Console, Vertical Section VPA-4 Digital data process

system, CRT display Pressure TAVG Narrow Range Temperature Recorder TAVG (Control) Narrow Range Temperature Indicator (540-610F)

Loop A, B,, and C

Wide Ran e

0-2300F Tem erature 8-Core Exit Thermocouples (2 per quadrant)

VPA-2 VPA-2 Subcooli n Mar ~in Wi'de Range (0-3000 psig) Pressure Indicator Narrow Range (0-1000 psig) Pressure Indicator Digital Indicator Filler Section Between VPA-11 and VPA-1 Sub'cooled Margin Monitor PRESSURIZER Pressure VPA-4 VPA-3 Narrow Ran e

1700-2500 si Recorder Pressurizer Indicator Channel I, II, and III

Level VPA-3 STEAM GENERATOR Pressure VPA-3 VPA-3 VPA-3 Level VPA-3 VPA-3 VPA-3 STEAM HEADER Pressure VPA-3 VPA-3 VPA-3

AUXILIARY FEEDWATER Flow

Console, Vertical Section - 3 Level 0-100K, s an Level Indi,cator Channel I, II, and III Pressure Indi cator 0-1200 si Steam. Generator A Channel I, II, and III Steam Generator B Channel I, II, and III Steam Generator C Channel I, II, and III Level Indicator Normal Ran e

0-1001 Steam Generator A

Ghannel I, II., and III Steam Generator B

Channel I, II, and III Steam Generator C

Channel I, II, and III Steam Header Pressure Indicator 0-1200 si Steam Header Channel I, II, and III Steam Header Channel I, II, and III Steam Header Channel I, II, and III Fl ow Indi cators 0-300 GPM Steam Generators A, 8, and C

HIGH'RESURE SAFETY INJECTION SYSTEM HPS'IS Pressure Header Pressure VPB-3 VP.B-3 Pressure Indicator (0-2000 psig)

Pressure Indi'cator (0-2000 psig')

Fl ow VPB-4 VPB-4 REFUELING WATER STORAGE TANK RWST VPB-3 CONTAINMENT SUMP LEVEL VPA-1 CONTAINMENT LEVEL (Minimum Level 2'.93 ft.

for ECCS recirculation phase)

VPB-4 CONTAINMENT PRESSURE VPB-2 VPB-2 CONTAINMENT RADIATION Process Radiation Header Flow Flow Indi cator (0-1000 GPM)

Flow Indicator (0-1000 GPM)

Level Indi cator 0-100K Level Indicator Level Recorder 0-100K, Level Recorder Level Indicator Red Indicating Light Illuminated. when 2.93 ft.

level is reached.

Narrow Range (-3".to +3") water gauge Wide Range (0-80 psig)

Radi oacti vit L'evel s 1-100,000 counts per minute Monitor Panel STEAM GENERATOR SECONDARY Radiation Detector SIDE RADIOACTIVITY

'Process Radi ati on Monitor Panel Liquid Sample CONDENSER RADIOACTIVITY Radi ati on Detector Process Radi ation Panel Steam Jet Air Ejecto r Outl et CONTAINMENT TEMPERATURE.

Tem erature Recorder 0-300F VPB-7 Containment Atmosphere

The required operator actions for the steam break accident and the loss of coolant accident are included in Turkey Point Unit No.

4 Emergency Operating Procedures.

The contents of the Emergency Operating Procedures are discussed during lecture sessions in training and requalification training programs.

During simulator training each operator is required to demonstrate proficiency duri ng simul ated emergency condi tions i nvol vi ng 1 oss of cool ant and steam break accidents.

STATE OF FLORIDA

)

)

COUNTY OF DADE

)

ss

~

John A. DeMastrl, being first, duly sworn, deposes and says:

Manager, Nuclear Licensing Light Company,,the herein; of'lorida Power That he has executed the foregoing document; that the state-ments made in this said document are true and correct to the best of his knowledge, information, and belief, and that he is authorized to execute the document on behalf of said John A. DeMastry Subscribed and sworn to before me this

./z 'ay of' 4

NOTARY=,PUBLIC, n and for the County of Dade,

,! State ~of~-Florida Notary Pubiic, Stato of Rorlda at Largo Ny commission expires:

\\t